-
Žana Popović (General Atomics)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
A number of advanced tungsten (W) based and chromium (Cr) plasma-facing materials were exposed to reactor-relevant divertor plasmas in the DIII-D tokamak using the Divertor Materials Evaluation System (DiMES). The objective was to evaluate the surface response and thermo-mechanical performance of next-generation divertor materials under steady-state and transient heat loads typical of future...
Go to contribution page -
Yevhen Zayachuk (UKAEA)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Following the third deuterium-tritium campaign, DTE3, and the end of JET operations in 2023, a selection of plasma-facing components was removed from JET vessel which included the modules of bulk tungsten divertor tile, so called “tile 5” or LBSRP. During JET operations, these bulk tungsten tiles were exposed to deuterium, tritium and helium plasmas. After the end of plasma operations and...
Go to contribution page -
Ladislas Vignitchouk (KTH Royal Institute of Technology)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
The extreme plasma heat loads arising during disruptions play a major role in determining the lifetime of plasma-facing components (PFCs). In particular, transient surface melting events are of crucial importance, not only because melt displacement constitutes a major PFC erosion mechanism, but also due to the risk of liquid metal filling the gaps between adjacent wall components [1]. In such...
Go to contribution page -
Shaun Haskey (Princeton Plasma Physics Laboratory)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Passive spectral measurements of Balmer-𝛼 are used to constrain DEGAS2 neutral transport simulations that show 20% of the neutral flux to the tiles is above 100eV, and 5% is above 1000eV at the outer strike point in typical DIII-D H-modes with pedestal top temperatures near 1000eV. Multistep charge exchange between ions and recycled neutrals transfers energy and momentum between the two...
Go to contribution page -
Estelle ROMULUS (IUSTI)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
In nuclear fusion, the extreme conditions inside of tokamaks, such as the WEST tokamak expose the plasma-facing components (PFCs) to intense heat and particles fluxes up to 10 MW/m². The interaction between the plasma and the PFCs is responsible for the temperature rise of these components, which are actively cooled to dissipate heat. Therefore, real-time temperature monitoring of the...
Go to contribution page -
Hugo Bufferand (CEA)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Liquid metals plasma facing component composed either of Lithium, Tin or Gallium are considered for heat exhaust in future fusion power plants. Different concepts from liquid metal in capillary structures or free flowing liquid metal layers have already been applied and considered for different devices [1,2] and are now investigated for future machines such as Renaissance Fusion stellarator...
Go to contribution page -
Corneliu Porosnicu (INFLPR)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
The performance, operational safety, and long-term viability of ITER’s magnetic confinement are determined by the interaction of hydrogen isotopes with plasma-facing components. As a full-tungsten reactor environment is now desired, boronization is essential for reducing impurity levels. Boron high affinity for residual gases such as oxygen and nitrogen, lead to improved plasma purity and...
Go to contribution page -
Francesco Cani (CNR-ISTP)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Plasma-wall interactions (PWI) are crucial in determining the overall performance and life-time of tokamak plasma-facing components (PFCs), particularly in high-performance machines like ITER and future fusion reactors. PWIs can lead to erosion, impurity generation, and fuel retention, negatively affecting plasma confinement and integrity of PFCs. Linear plasma devices, such as GyM [1] and its...
Go to contribution page -
Steven Thériault (University of Toronto)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Plasma facing components in fusion devices are subjected to very high heat fluxes, leading to the potential for material degradation and melting. In the case where PFCs are made out of high Z materials like tungsten, the subsequent contamination of the plasma from wall degradation can become a major problem leading to a shutdown of the fusion reactions. To prevent the heat fluxes from damaging...
Go to contribution page -
Olga Ogorodnikova (Moscow Engineering Physics Institute)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The fuel in thermonuclear and hybrid fission/fusion reactors will be a deuterium-tritium (DT) mixture. Plasma-facing materials will be exposed to both stationary and high heat pulsed plasma loads, as well as helium and neutrons, arising from the DT reaction. Such extreme operating conditions can lead to degradation of materials due to erosion, formation of plasma- and radiation- induced...
Go to contribution page -
Dazheng Li (FZJ-IFN1)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The penetration depth of pellets mainly depends on the ablation rate, which further influences fueling efficiency. In this work, based on the neutral gas shielding model [1], we proposed a neutral gas evaporation shielding (NGES) model, in which the ablation cloud radius was evaluated self-consistently to account for the real-time shielding effect of ablated gas on the incoming heat flux. To...
Go to contribution page -
Joey Demiane (MIT - PSFC)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Understanding tritium (T) and deuterium (D) retention in boron-coated tungsten (B/W) plasma-facing components is essential for next-generation fusion devices such as SPARC, ITER, and future commercial reactors, because wall retention sets the in-vessel T inventory and influences D/T recycling and wall pumping, with direct implications for safe operation and sustained plasma performance. In...
Go to contribution page -
Kentaro Masuta (Kyushu University)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
In the realization of future fusion reactors, the strict management of tritium inventory and the minimization of tritium permeation from plasma-facing components (PFCs) into the coolant systems are critical issues. These challenges directly impact both the radiological safety of the power plant and the fuel efficiency of the fusion cycle. Tungsten (W) is currently the leading candidate...
Go to contribution page -
Minyou Ye (University of Science and Technology of China)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The integrity of plasma-facing components (PFCs), particularly tungsten (W), against synergistic irradiation damage, specifically from high-energy particles, high-flux plasmas, and high heat loads, remains a critical challenge for future magnetic confinement fusion reactors. This research investigated the combined effects of high-flux, low-energy plasma, and high-energy particle beams on...
Go to contribution page -
Zhe Liu (University of Science and Technology of China)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Tungsten (W) is the leading candidate material for plasma-facing components (PFCs) in future magnetic confinement fusion reactors, such as ITER and DEMO, due to its excellent properties. The PFCs will face extreme operating environments, including intense fluxes of high-energy particles, plasmas and heat, which will cause significant radiation damage and affect fuel retention.
Go to contribution page
This research... -
Sabina Markelj (Jožef Stefan Institute)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
It was shown experimentally for tungsten, that He retained close to the surface influences transport and retention of hydrogen isotopes (HI). Namely, experiments using He seeded D plasmas, showed that the addition of He leads to reduced blistering accompanied by reduced D retention [1].
Go to contribution page
Recently, we have performed a systematic series of D exposures in tungsten where He was pre-implanted near... -
Mr Jonathan DUFOUR (CEA)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Plasma-wall interactions in tokamaks lead to transport and retention of hydrogen isotopes, especially tritium, in the plasma exposed material causing safety issues. Cleaning procedures are envisaged in ITER to recover the tritium [1] using pure deuterium (D) operation to trigger isotope exchange. This work presents isotope exchange simulations and experimental validation of the model, which...
Go to contribution page -
Thomas Schwarz-Selinger (MPPL)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
EUROFER97 is the European Reduced Activation Ferritic-Martensitic (RAFM) candidate steel to be used as a structural material in future nuclear fusion devices. Neutron irradiation will degrade the mechanical properties, setting limits in terms of operational temperature and maximum allowed dose. It is anticipated that the European DEMO will utilize a first blanket with a 20 dpa damage limit in...
Go to contribution page -
Prof. Sergei Krasheninnikov (UCSD)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Even though the importance of the mesoscale structures (e.g. “avalanches”, “blobs”, vortices) in
Go to contribution page
the magnetized plasma transport was recognized a long time ago (e.g. see [1-4]), still the physics
of such objects has many open issues and the extensive studies of these phenomena, both
theoretical and experimental, are continue (e.g. see [5] and the references herein).
The avalanche is... -
Mr Jae-in Song (KFE)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
This paper presents the design and construction of a Divertor Fueling System (DFS) developed for the advanced KSTAR operation under diverted discharge configurations aimed at enabling mainly high-density plasma operation and detachment control. The fueling system essential for plasma initiation, density regulation, radifo-frequency (RF) heating coupling, vacuum-wall impurity flushing,...
Go to contribution page -
June-Woo Juhn (Korea Institute of Fusion Energy)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The KSTAR tokamak is now under transition period of the plasma facing materials from carbon to tungsten: the carbon tiles as the bottom divertors are replaced with tungsten mono-blocks before 2023. The remaining PFCs such as passive stabilizers, top divertors and limiters will be coated tungsten on the existing carbon tiles for the upcoming 2027 experimental campaign. Accordingly, the normal...
Go to contribution page -
Sebastian Rode (Forschungszentrum Jülich GmbH, Institute of Fusion Energy and Nuclear Waste Management – Plasma Physics, 52425 Jülich, Germany)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The tritium fuel cycle is an important aspect for the design of a fusion reactor, since the initial on-site amount of tritium must ensure self-sufficient reactor operation with required fusion power and tritium breeding efficiency, complying at the same time with safety limits on in-vessel tritium accumulation. Retention of hydrogenic species inside plasma-facing components (PFCs) represents a...
Go to contribution page -
zeshi gao (University of Science and Technology of China)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The transport and retention of hydrogen isotopes is of vital importance for the realization of future commercial fusion reactors because it is closely related to plasma operation, fuel recycling, and radiation safety. Tungsten (W) is a leading plasma-facing material, and its performance can be significantly enhanced by the incorporation of small amounts of ultrafine oxide particles (such as...
Go to contribution page -
Bianca Solomonea (National Institute for Laser, Plasma and Radiation Physics, Bucharest, Romania, Doctoral School of Physics, Faculty of Physics, University of Bucharest, Magurele-Ilfov, Romania)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Boronization plays a fundamental role in impurity control and plasma conditioning in fusion devices. With tungsten now implemented as the primary plasma-facing material in ITER, new constraints have emerged regarding impurity transport, plasma stability, and hydrogen isotope retention. One of the main motivations for using boron is the high affinity for oxygen and other impurities, a...
Go to contribution page -
Dr Yuri IGITKHANOV (KIT)21/05/2026, 15:55C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Abstract
Go to contribution page
The fusion power plant operation with a fuel cycle including a direct internal recycling aims to increase the fuel burnup fraction and to reduce the tritium inventory. The expected accumulation of protium and fuel imbalance requires control of plasma fueling and particle exhaust. The formation of protium and tritium in the reactor chamber in concomitant to DT fusion reactions and... -
Suguru Masuzaki (National Institute for Fusion Science)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
In the Large Helical Device (LHD), boron powder injection experiments have been conducted using an impurity powder dropper (IPD), developed by PPPL, for real-time wall conditioning. A pronounced effect of boron powder injection has been observed on the line emission intensity of iron, the main constituent element of the stainless steel type 316L first wall material in LHD. The emission...
Go to contribution page -
Raymond Diab (Massachusetts Institute of Technology)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
When the density at the Ion Cyclotron Range of Frequencies (ICRF) antenna limiter’s edge falls below the lower hybrid (LH) resonance density (S = 0 in Stix dielectric tensor), the slow wave (SW) can propagate in front of the antenna. This wave carries large parallel electric fields that can strongly enhance the sheath potentials on the antenna limiters, thereby increasing the sputtering yield...
Go to contribution page -
Riccardo Ian Morgan (SPC-EPFL)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
In tokamaks, intense heat fluxes strike the divertor targets, risking damage and core plasma contamination by eroded wall material. Future reactors must therefore operate in a detached regime, where heat loads and plasma temperatures near the walls are greatly reduced [1]. This can be achieved by impurity seeding, which promotes radiative cooling and momentum losses in the boundary plasma....
Go to contribution page -
Naren Varadarajan (CEA)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
The WEST tokamak has been used to study long-pulse scenarios in a full Tungsten environment. Its contamination in the core, which reduces performance, as well as the location of erosion/deposition sites are important. It continues to be difficult however, to get accurate estimates of sputtered fluxes and Tungsten content in any given pulse. Modelling that aims to answer this question should...
Go to contribution page -
Alexis Huart (CEA)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
A major issue for next step devices is the control of plasma wall interaction, both for keeping the material erosion compatible sufficient lifetime of the components as well as for mitigating core contamination by high Z impurities and consequently the reduction of plasma performances. In this respect, WEST experiments supported by numerical modeling are particularly relevant to progress in...
Go to contribution page -
Timo Dittmar (FZJ)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
The complex 3D magnetic topology of the island divertor in Wendelstein 7-X (W7-X) produces feature rich erosion and deposition patterns on the divertor plates and strongly affects impurity transport in the edge plasma region. While crucial for stellarator optimization towards reactor-relevant configurations, modelling impurity transport is challenging and requires thorough code validation....
Go to contribution page -
Zhuang Liu (Soochow University)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
ITER has switched to full tungsten wall configuration[1].High-Z tungsten dust originated from the Plasma-Surface Interactions(PSIs) may result in the degradation of plasma discharges, the H-L mode back transition or even disruption. Meanwhile, low-Z impurity pellets such as lithium or boron ones are strong candidates for ELM control with impurity injection. The small or grassy ELMs are...
Go to contribution page -
Tian Xie (Northeast Agricultural University)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
Nitrogen with the high chemical activity can generate the complex chemical compounds with other elements (i.e. hydrogen, deuterium) in the plasma volume or in the divertor target material. These chemical processes of nitrogen impurity cannot be simulated in the modelling, which is attributed to the limited version of the current EMC3-EIRENE code. Nevertheless, these chemical effects can be...
Go to contribution page -
Juhyeok Jang (Korea Institute of Fusion Energy)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
Tungsten impurities in tokamak plasmas introduce significant challenges due to their high radiative efficiency. Following the recent update of the ITER baseline, the importance of understanding tungsten impurity generation and transport has increased significantly. In KSTAR, the installation of a tungsten monoblock cassette divertor in 2023 has opened a new operational regime for ITER-relevant...
Go to contribution page -
Atsushi Okamoto (Nagoya Univ.)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
Experimental investigation of impurity transport in the scrape-off-layer (SOL) and divertor region of tokamaks is crucial for developing accurate impurity transport models. Since various divertor configurations are proposed for future fusion reactor, flexibility in magnetic field configuration is required for experimental devices. Some tokamaks provide various divertor configurations such as...
Go to contribution page -
Liang Liu (Southwestern Institute of Physics)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
Plasma-wall interaction (PWI) during steady phases and transient events such as disruptions and edge localized modes (ELMs) is a critical issue in the magnetically confined plasmas, mainly because the particles and heat fluxes from plasmas need to be controlled to prevent damage to the plasma facing components (PFCs), as well as the impact of impurity influx on the main plasma performance...
Go to contribution page -
Sven Wiesen (DIFFER - Dutch Institute for Fundamental Energy Research, De Zaale 20, 5612 AJ Eindhoven, Netherlands)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Reactor-scale divertor design demands rapid yet physics-consistent exploration of exhaust operational spaces. Fusion power plants (FPPs) beyond ITER require fully integrated scenario development that ensures both robust core-plasma performance and compliance with engineering limits on plasma-facing components (PFCs), particularly with respect to heat fluxes and wall erosion. Central to this...
Go to contribution page -
Charles Hirst (University of Wisconsin-Madison)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Fabrication of first wall components for Fusion Pilot Plants (FPPs) through conventional powder metallurgy routes may be unfeasible due to cost and durability concerns. Instead, technologies such as additive manufacturing could be used to produce near net shape components with specifically-tailored microstructures. Cold spray deposition is another strategy that can be used to produce thick...
Go to contribution page -
Annemarie Kärcher (Max Planck Institute for Plasma Physics, 85748 Garching, Germany (IPP), Technical University of Munich, Germany (TUM))21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Ab-initio modeling of helium (He) behavior in tungsten (W) predicts clustering mechanisms, self-trapping and trap mutation, but to date there is little experimental evidence for isolated He self-trapping in a defect-free lattice. In order to investigate these mechanisms and the influence of intrinsic defects in W on them, the effects of He implantation into polycrystalline (poly-W) and...
Go to contribution page -
Kelly Garcia (Universität Greifswald)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
The island divertor concept in Wendelstein 7-X (W7-X) is the leading candidate for future stellarator reactors for handling heat and particle exhaust [1]. However, to date the study of this concept has primarily focused on experimental scale and shown that achieving detachment and high-recycling is more complex than standard tokamak poloidal divertors due to the complex 3D magnetic geometry....
Go to contribution page -
weixi chen (National Institutes for Quantum Science and Technology)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
A limiter is a plasma-facing component in a fusion reactor that protects blanket modules from excessive surface heat load and high-energy particles interaction. It achieves this by protruding from the first wall (FW), effectively shadowing the blanket. The limiter extends continuously in the poloidal direction, except near the divertor, and is installed in each 90-degree toroidal segment. To...
Go to contribution page -
Marcos Navarro (University of Wisconsin-Madison)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
EMC3-EIRENE modeling of detachment of 3D scenarios at DIII-D with resonant magnetic perturbations applied show that neon as a seeded impurity is able to semi-detach the ITER similar shape at comparatively lower separatrix impurity densities than seeded nitrogen. This impurity buildup is observed on the low-field side at the DIII-D shelf, leading to detachment of the far scrape-off layer (SOL),...
Go to contribution page -
Dylan Kohler (University of Wisconsin-Madison)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Fabricating fusion pilot plant (FPP) first-wall components through additive manufacturing (AM) could enable cost-effective, scalable production of complex, functionally graded components. In addition, optimized AM processes can produce materials that retain, and in some cases improve, the key thermo-mechanical properties needed in an FPP, relative to conventionally processed tungsten.
To...
Go to contribution page -
Prof. Zongxiao Guo (GNOI)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Following the decision by ITER in its new baseline to replace beryllium with tungsten in the first wall, the fabrication of large-area curved plasma-facing tungsten components has attracted growing attention. Low-cost, high-efficiency manufacturing of tungsten (W) first wall structures for breeding blankets holds significant importance for future large-scale tokamak fusion experimental...
Go to contribution page -
Derek Harting (FZJ - Forschungszentrum Jülich GmbH)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
With the transition to a full-tungsten (W) machine, detailed understanding of the W erosion source and transport processes are key to assessments of the main plasma performance. In work conducted under the auspices of the ITER Scientist Fellow Network, the capability of the recently upgraded kinetic ion transport model (KIT) of EMC3-EIRENE [1] for W transport with the full shaped 3D wall...
Go to contribution page -
Marlene Patino (University of California, San Diego)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Previous work utilizing deuterium (D) found that while displacement damage increases D retention in tungsten (W) due to generated D trapping sites [1-3], helium (He) pre-treatment or seeding reduces D retention in pristine W for low-energy D since He nanobubbles act as a diffusion barrier to D [2,4]. He pre-treatment or seeding also reduces D retention in pre-damaged W such that under certain...
Go to contribution page -
Hans Maier (MPI for Plasma Physics)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
For the ITER plasma-facing first wall the original plan of using beryllium as a wall material has been changed to using tungsten. In this configuration the first operational phase, will be performed with an inertially cooled “Temporary First Wall” [1]. The elements of this wall will experience different loading conditions depending on their respective locations. Correspondingly, the design...
Go to contribution page -
Yuta KINASHI (Plasma Reseach Center, University of Tsukuba)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Linear plasma devices have widely been utilized to examine plasma-facing materials under fusion reactor relevant conditions. In addition to DC arc sources, helicon plasma-based devices have also been developed. However, impurity generation from dielectric vacuum windows surrounded by an RF antenna remains a critical issue, as observed in Proto-MPEX [1], since unwanted deposited impurities can...
Go to contribution page -
Jeremy Lore (Oak Ridge National Laboratory)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Liquid metal (LM) divertor conditions for a pilot-plant class tokamak are simulated using a multi-component integrated model which considers plasma transport, LM magnetohydrodynamics and heat transfer, and formation and transport of lithium released from the divertor surface. Fast-flow LM divertors have several intrinsic advantages over solid surface concepts due to their self-replenishing...
Go to contribution page -
Matthieu Spinosi (Institut Jean Lamour IJL UMR CNRS 7198, Université de Lorraine, 54011 Nancy, France)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Keywords: Magnetic confinement Fusion, liquid metal, capillary structures, lattice, tungsten, additive manufacturing
One of the main challenges in the development of a magnetic confinement fusion reactor is the durability and maintenance of internal components, particularly the divertor, which is exposed to the highest heat and particle fluxes. To withstand these extreme conditions,...
Go to contribution page -
Dr Nobuyuki Asakura (National Institutes for Quantum Science and Technology (QST))21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
SONIC divertor code has simulated the divertor performance of power exhaust by seeding impurity (Ar) and He exhaust in the detached divertor for JA DEMO design (1.5 GW-level fusion power), where exhaust power, fuel and He particles at the core-edge boundary of 250 MW, 1x$10^{22}$ D/s and 5.3x$10^{20}$ He/s, respectively, were given [1]. Recently, NEUT2D and IMPMC codes (kinetic MC modellings...
Go to contribution page -
Daniel Alegre Castro (Laboratorio Nacional de Fusion. CIEMAT. Madrid)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Tungsten is the preferred candidate as plasma-facing material for future power plants like DEMO. However, current W materials produced by industry like ITER-Grade Tungsten (IGW) will most likely not meet their requirements. New advanced armor concepts are being studied to withstand those steady state and transients heat loads [1]. One of the methods to simulate the heats loads expected in a...
Go to contribution page -
Marianne Richou (CEA)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Since the introduction of the new ITER baseline [1], significant efforts were made in selecting the Tungsten (W) based armour materials for the inertially cooled Temporary First Wall (TFW). The TFW provides an environment for ITER’s initial operation campaign “Start of Research Operations (SRO)” that is more forgiving to disruptions while mimicking the final actively cooled First Wall in terms...
Go to contribution page -
Dr Vlad Soukhanovskii (Lawrence Livermore National Laboratory)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Managing extreme heat and particle fluxes to divertor targets remains a major challenge in spherical tokamaks. In this work, we use UEDGE simulations to investigate three interconnected topics relevant to NSTX-U operation: (1) graphite plasma-facing components (PFCs) performance, (2) lithium PFCs vapor shielding and its dependence on upstream plasma conditions, and (3) the impact of snowflake...
Go to contribution page -
Erik Wüst (Forschungszentrum Jülich GmbH, Institute of Fusion Energy and Nuclear Waste Management – Plasma Physics)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
Laser-induced Breakdown Spectroscopy (LIBS) is a method for the analysis of material composition and is used for Plasma-Wall Interaction studies. Especially for the detection of implanted fuel, deuterium (D) and tritium (T), in plasma-facing components (PFCs) in a fusion reactor, its in-situ capabilities are promising. In the case of tungsten (W) PFCs, the expected low levels of implanted fuel...
Go to contribution page -
Dr Zori Harutyunyan (Imperial College London,London, UK)21/05/2026, 15:55J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsPoster
To improve the performance of plasma-facing components, various advanced concepts of W materials are being developed. In particular, tungsten-based refractory high-entropy alloys (HEAs) in the W–Ta–V–Cr–Ti system are attracting attention due to their excellent radiation resistance and strength at high temperatures. In this work, the intentional addition of boron to the ...
Go to contribution page -
Holger Reimerdes (École Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), Lausanne, Switzerland)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The Swiss Plasma Center is upgrading the TCV tokamak to test a tightly baffled, long-legged divertor (TBLLD), a novel concept that promises to enhance power exhaust capabilities with minimal modification to the magnetic configuration [1,2].
Go to contribution page
Simulations using the SOLPS-ITER code indicate that a TBLLD can improve TCV’s power exhaust capability by an order of magnitude compared to the unbaffled... -
Jack Lovell (Oak Ridge National Lab)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
We report on the development of the X divertor configuration [1] on MAST Upgrade and the reduction in heat and particle fluxes at the divertor targets this configuration achieves over a conventional divertor configuration. Using the Toksys [2] and TED [3] frameworks, we design magnetic equilibria with increased poloidal flux expansion at the target compared with a conventional divertor...
Go to contribution page -
Stijn Kobussen (DIFFER, Dutch Institute for Fundamental Energy Research)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The safe and controlled exhaust of heat from magnetically confined fusion plasmas requires having dynamic information about the system. On present-day reactors, this information is obtained using system identification experiments, which involve perturbing the gas injection rate and observing the Scrape-Off-Layer (SOL) response in the frequency domain. However, this strategy is cumbersome for...
Go to contribution page -
Felix Albrecht (Max Planck Institute for Plasma Physics, Garching, Germany and Technical University of Munich, TUM School of Natural Sciences, Physics Department, 85748 Garching, Germany)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
In a future fusion reactor, the power crossing the separatrix must be at least 100$\,$MW [1], with a power fall-off length $\lambda_q$ in the order of millimeters [2]. This presents a significant challenge to the divertor target plates, which cannot be solved without strong radiative losses in the divertor volume, leading to detachment. Thus, high impurity concentrations are required in the...
Go to contribution page -
Martim Zurita (EPFL - SPC)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Mitigating the heat load of type-I edge-localised modes (ELMs) via impurity seeding has been investigated in single-null plasmas of the Tokamak à Configuration Variable (TCV). The measurements exploit TCV’s fast diagnostics to resolve ELMs, from the plasma edge to the divertor targets: the ELM energy loss is measured with a diamagnetic loop (DML, acquisition frequency of f=10kHz); the...
Go to contribution page -
Sid Leigh (UKAEA)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
We present the modelling of detachment burn-through scenarios with a new exhaust model developed in ReMKiT1D [1], a framework for 1D fluid-kinetic modelling and collisional-radiative interactions, and the 2D fluid modelling suite SOLPS-ITER. Starting from a "1D" SOLPS case, a detached plasma background representing the MAST-Upgrade Super-X divertor is reproduced within the ReMKiT1D exhaust...
Go to contribution page -
Dr Juuso Karhunen (VTT)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Introduction of cross-field drifts in SOLPS-ITER [1] simulations of STEP [2, 3] with fully tracked Ar impurities was found to re-distribute the impurities in the simulation volume differently from the main ions, improving the representation of Ar in the plasma from the first drift simulations for STEP [4], where Ar was proxied as a fixed fraction of the main ion content, and complementing...
Go to contribution page -
Koyo Munechika (ITER Organization)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Accurate assessment of radiated power during Shattered Pellet Injection (SPI)–mitigated disruptions is essential for ensuring the protection of ITER plasma-facing components [1].
Go to contribution page
In this work, we extend the synthetic bolometry framework for ITER by incorporating a full 3D treatment of the diagnostic geometry, including inner apertures and sub-collimators, using the CHERAB framework.
This... -
Fabian Solfronk (IPP)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Efficient power exhaust is a critical challenge for fusion power plants. The so‑called X‑point radiator (XPR) has been proposed as a promising concept to redistribute and dissipate heat loads via impurity radiation.
While detailed edge‑plasma simulations provide valuable insights, they are computationally intensive and therefore not ideally suited for rapid design iteration or extensive...
Go to contribution page -
Andreas Redl (Commonwealth Fusion Systems)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The SPARC tokamak, a high field (12 T), high current (8.7 MA) machine designed to achieve an energy gain Q of 11 in H-mode with DT fuel, is currently under assembly and will be expected to be in operation in 2027. As the expected heat-flux width for SPARC is in the range between 0.3 and 0.6 mm, the power exhaust is extremely difficult due to upstream steady-state parallel heat fluxes of about...
Go to contribution page -
Anastasios Tsikouras (MPPL)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The high heat fluxes in magnetic fusion devices pose an immediate threat to their plasma facing components (PFCs). These fluxes can be mitigated by injecting low to medium Z impurities [1]. These impurities stimulate radiation emission in the edge of the plasma, which dissipates power volumetrically and reduce the heat fluxes to the PFCs. Efficient control of the radiated power ($P_{rad}$)...
Go to contribution page -
Eva Havlickova (IRFM, CEA Cadarache, France)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Until lately, codes such as SOLEDGE3X [1] or SOLPS [2] have been routinely used without including drift effects, although past studies have shown their significance [3,4]. Thanks to the continuous development and improvements in the numerical treatment, the inclusion of drifts has now become more common and proves to be important in detachment studies [5,6]. In this contribution, we evaluate...
Go to contribution page -
Amit Kohinoor Kharwandikar (MPPL)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The island divertor is the leading plasma exhaust concept in stellarators, and evaluating its reactor relevance is a key objective of the Wendelstein 7-X (W7-X) experiment [1]. From a power exhaust perspective, maintaining steady-state divertor heat loads below 10 MW/m² is essential for safe and sustained device operation. However, a robust framework for quantitatively characterizing the 3D...
Go to contribution page -
Aaro Järvinen (VTT)21/05/2026, 15:55G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The design of the ITER divertor was established via guidance of extensive scoping studies of the baseline burning conditions at $Q_\text{DT} = 10$, conducted with the SOLPS-4.3 boundary code without cross-field drifts or currents [1, 2]. Since those scoping studies, the development of the SOLPS-ITER code package, launched by the ITER Organization in 2015, has enabled numerically robust and...
Go to contribution page -
Cas Robben (DIFFER)21/05/2026, 15:55B. Material Erosion, Migration, Mixing, and Dust FormationPoster
The extreme conditions in ITER demand a carefully selected first wall material. Tungsten, which replaced the originally planned beryllium, offers improved resilience and reactor relevance but introduces new challenges in impurity management. A thin boron coating (~100 nm) has been proposed to mitigate these issues [1]. However, during ITER operation, erosion and redeposition processes are...
Go to contribution page -
Mr Romain Avril (Université de Lorraine, Institut Jean Lamour, UMR 7198 CNRS)21/05/2026, 15:55E. Impurity Sources, Transport and ControlPoster
Recent developments in advanced nuclear fusion reactors consider the use of solid plasma-facing components (PFCs), typically made of tungsten. However the emergence of new compact nuclear fusion reactor concepts, presented as more viable for commercial applications, can lead, due to the reduced plasma wetted surface, to heat fluxes much higher than the 15MW/m$^2$ estimated for the ITER...
Go to contribution page
Choose timezone
Your profile timezone: